ML20042E539

Forwards Response to Questions Re High Enriched U/Low Enriched U Conversion,Per NRC 890710 Ltr

ML20042E539
Person / Time
Site:	05000199
Issue date:	08/10/1989
From:	Berlin R MANHATTAN COLLEGE, RIVERDALE, NY
To:	Michaels T Office of Nuclear Reactor Regulation
References
NUDOCS 9004230560
Download: ML20042E539 (26)

Text

F,. ' m p^- MANMATTAN oOLLEGE PANKWAY 4

'enheMan RfVERDALE. NEw YORK 10ert WCCHANICAL ENQiNiERING DEPARTMENT '

. . (212) 980 4146 August 10,1989

. Theodore S. Michaels, Project Manager

' Standardization and Non Power Reactor Project Directorate Division of Reactor Projects lit,lY, Y, and Special Projcets Office of Nuclear Reactor Regulations U.S. Nuclear Regulatory Commission Washington, DC 20$$$

SUBJECT:

Response to Questions Regarding !!EU/ LEU Conversion at Manhattan College (your letter of July 10, 1989)

Dear Mr. Michaels:

Enclosed are the responses to questions raised in your letter of July 10,1989 relative to the Safety Analysis Report we submitted regarding the llEU to LEU fuel conversion for the Manhattan College Zero Power Reactor. Please let me know if any additional information is required.

Sincerely, s'fae k Robert E. Berlin Reactor Administrator f Nf Y gg, sa. tmsas t i

,oo m osso e90eto p pn nwcx oseg

MANilATTAN COLLEGE ZERO POWER REACTOR .

QUESTIONS AND ANSWERS (1) Q What is the calculated 'just critical' mass? j A 'Just Critical U 235 Mass"in the LEU core is calculated by the Argonne National Laboratory (ANL) as 235 x l$ + 27.4 x 1 = 3$$2,4 grams (total U 235), where 15 represents the number of full fuel elements and I represents partial fuel elements.

(2) Q What is the fuel element worth versus grid position for the LEU fuel? ,

A There was no calculated "Fuct element worth versus grid position" provided by the ANL for the LEU fuel, since there was no such information provided by AMF Atomics for the IIEU fuel in 1965 for comparison purpose, liowever, detailed calculations on two control rods (I regulating rod and I shim rod) have  ;

been made by the ANL, using both the Monte Carlo method and diff usion theory (*)-  !

(3) Q Are there provisions for any out of core fuel storage? Please explain what provisions have been made to safely store the IIEU fuel elements in the event .

that shipment is not possible on the day they are removed from the core. l A There is onsite capability for storage of all the IIEU clcments af ter removal from the reactor. The elements would be placed individually into cylindrical sleeves, and then placed four to a container in the original SYLCOR shipping containers the fuel elements were received in, This temporary storage procedure has been -

used in the past during tank cleaning and maintenance, and is documented in r MCZPR records and in the August,1983 SAR. ,

(4) Q Please provide any lieu versus LEU comparisons of power distributions in the fuel elements, and any power distribution versus fuel loading information in the partially loaded element.

A The power distribution and nuclear power peaking factors that were calculated 5 by the ANL for the existing IIEU core and the LEU reference core with the shim and regulating rods fully withdrawn are shown in Figure 1. (*) The power distributions show the power per fuel element (in milliwatts) and the power peaking factors show the absolute peak power density in each fuel clament (computed at the edge of the mesh interval with the highest power) divided by the average power density in the core fuel.

The data in Figure I shows that the power distributions and total power peaking factors are nearly the same in the lieu and LEU cores. Ilowever, the limiting

• fuel cicment in the lieu core is located in grid position 33 and the limiting fuel  !

element in the LEU core is located in grid position 34. This is because the location of one fuel element was changed in the LEU core (from position 46 to position

14) to increase the reactivity worth of the regulating rod.

Instead of power distribution versus partial fuel loading in the LEU reference ~

core, ANL provided us with the changes of excess reactivity due to the presence '

of the partial fuel element, as shown in Table 1 (*)

(a) 3.E. Matos and K.E. Frcssc,

• Analyses For Conversion of the Manhattan College Zero Power Reactor from HEU to LEU fuel". ANL, February,1989. ,

1 u

% 4 (5) Q Neglecting bias, calculational uncertainty, etc., what are the Manhntinn (best estimate) published values for regulatory and shim rod worth? Please give a technical justification for the values you choose. In reference to Table 3.2, page 19, please explain the meaning and operational implications of the two paren.

thetical statements about ' biases".

A Our Manhattan published values for regulating rod worth is -0.9%Ak/k , and for shim rod worth is-2.5%Ak/k. These values were measured by AMF Atomics for a critical assembly of the PTR reactor which has the same core as our MCZPR.

'Blases' here means the deviations of rod worth between previously measured values (by AMF Atomics) and current simulation data (by ANL) on the same lieu core. The major causes of such rod worth biases are:

(a) methods used for calculation, such as the Monte Carlo method and diffusion theory, (b) U 235 fissile loading variation (generally *2% ), and (c) sensitivity caused by ppm Boron equivalents in fuel plate cladding materials.

Since ANL has included all the *blases" possibilities in their analyses for the LEU reference core, safe operation can be expected as long as rod integrity is-maintained.

(6) Q Picase resubmit page 19 to show the deltas in Table 3.2 that are missing (% Ak/k) and correcting an apparent typo in section 3.3.1, seventh line, viz. reversing vs.

revising.

A These corrections have been made (see enclosure).

(7) Q ... Please provide NRC with the details of your Zero Power Physics Test program in this area. Additionally, ... you should prepare a fuel loading plan ... Please provide such a plan.

A A: MCZPR Physics Test Progrnm on I EU Core Excern Renetivity Part 1: Normal Tap): Water Temperaturc Reactivity Tests (60-80' F)

IIEU core excess reactivity measurce. cat v!u!cr normal tank temperature has been a routine experiment of the MCZPR operational program since 1965 (see Attachment I). Although reactivity shows a slightly positive value in the current ilEU core, it is much lower than the allowable peaking value of 0.44%Ak/k.

The same experimental procedures will be followed for the LEU core excess reactivity test; however, it will be conducted each time with an increasing S*F step wise temperature change to ensure that maximum excess reactivity of 0.44%Ak/k will not be execeded within the range of 60-80' F.

Part II: liigh Tank Water Temperature Reactivity Tests (80* F and up)

Since maximum excess reactivity has been measured (not calculated) at 110.6* F for the llEU core, the same result may also occur during isothermal heating (based -

on ANL analysis). For safety reasons, we will repcat the reactivity test at each 3- S'F temperature increment a few times during a period of severaldays running, with the same procedure as in Part I to insure l n) uniform temperature in the tank, i

1.

. l, b) consistent reactivity in the core, and c) good agreement with ANL calculated data, if measured excess reactivity of 0.44%Ak/kis reached well below a temperature of Ilo.6* F we will stop the tests and work with ANL to obtain further analysis and subsequent verification before our Technical Specifications revisions.

A B: Liel Loading Part 1: Fuel Loading Plan llEU fuel element removal and insertion has been performed at least every six years since 1965 for tank cleaning and component maintenance purposes. In order to avoid the abrupt changes of core reactivity as well as power level, we had proposed the fuel removal and insertion sequence, as shown in Table II, based on data recorded during the past 25 years. Table 11 shows that console meter readings were made during each 2 3 fuel element removal and insertion periods, for reactor power level (same as neutron multiplication) and Gamma radiation level checks. These standard procedurcs and the previous recorded data will be used as reference for 1 LEU / LEU core conversion.

Part II: Fuel Loading Beyond Criticality Based on " Reactor Period and Reactivity" experiment in the MCZPR (see Attachment I), reactor transient power has to be temporarily 25% over critical power in order to avoid an involuntary scram. This 125% power level (0.125 watt) was approved by the NRC during our license renewal in 1985.

As shown in Table 31 of Revision 4 of our Technical Specifications (see page 6), that both liigh Neutron Flux " count rate channel setting" and " linear channel setting" are allowed to reach.125% of full power rating. These safety systems (Table 31), which are particularly designed for reactor transient analysis would allow us to handle LEU fuel loading beyond criticality.

A C: Renetor Power Level Determlnntion and Rod Worth Calibrntion

" Power Level Determination" and " Rod Worth Measurement"(includes Rod Worth Calibration processes) have been two important routine experiments in the MCZPR since 1965 (sec Attachment II). Two methods will be employed for LEU core analysis in each of these two experiments. Recorded data from previous llEU core experiments will be used as reference for LEU core measurements.

(8) Q In the current LEU SAR, it is assumed that handling accidents where the fuel element is dropped results in no clad breach, is this scenario the same as the handling accident discussed in your 1983 IIEU SAR supporting the relicense?

A Yes, it is.

(9) Q Please explain how you will preparc the emergency shutdown rod for rapid use in the event of an error or other occurrence during fuel changes. (See section 4.2, S A R).

A The manual B C4 emergency shutdown rod, which is capable of shutting down the reactor in itself is located on the wall of the reactor facility within arms reach -

of a person standing on the platform, it can be rapidly lif ted from its supports and manually inserted in the reactor core to accomplish immediate shutdown.

4

l.

(10) Q Please resubmit only the revised sections of the Technical Specifications needed to accommodate the HEU to LEU conversion and any other change that you plan to make Provide a brief rationale for the changes.

A The proposed revisions to the Technical Specifications and rationale are provided in Attachment III. This will supercede the proposed revisions 5 and 6 included in our May 8,1989 submittal.

i 1

v - - . . ~ ~ .v < a

those which would appear in the proposed Technical Specifications for the LEU core (Rev.6), are listed in table 3.2. The technical specifications for the LEU Core (Revision

6) are included as Attachment til to this Report.

TABLE 3.2 Parameters of IIEU and LEU cores Core Page No. in Renetor Parameters liLU LEU. Tech Spec (rev. 6)

Exects Reactivity, % Ak/k 0.32 0.40 1.1 +0.4 ' 31 (with -1.0% Ak/k Dias to LEU Core)

Worth of Reg. Rod, % Ak/k (with -0.9 1.3 32

+0.3% Ak/A Blas to LEU Core)

Shutdown Margin, % A A/k 0.5 0.6 31 (with Shim Rod Stuck Out)

Worth of Shim Rod, % Ak/k 2.5 +3.4 32 3.3 DescrIntion of Fuel Removal and Pentnecment 3.3.1 Stens in Removal and Reolneement Processes During the process of l{EU/ LEU core conversion, each IIEU fuel element will be removed from the core and lowered into the fuel container (fuct cask) supplied by the EG&G Co. A sufficient number of containers will be obtained such that all 16 fuct elements (15 full and I partial fuct elements) can be sequentially removed from the core at one time, and then shipped to the DOE repository site, immediately after the completion of IIEU fuel removal, the new LEU fuct will be installed into the core, reversing the procedure for ilEU fuci element removal. In order to avoid an abrupt change of reactivity in the reactor core and to prevent the fuel cicments from obstructing each other, all the outer (circumferential) llEU fuel elements will be removed prior to that of the central elements and a reverse process of installing the LEU fuel elements will be carried out from the center of the reactor core. The detailed removal r.nd replacement sequence are schematically shown in Table 3.3 and Figure 3 2.

I 19

. I

c .

s -

'- Figure 1. Power Distributions and Power Penking Factors i' l

llEU CORE f Power /Eleinent. 5.4 -

6.2 5.2

• milllWalte I s2 I I 43 I I s4 i 1

Peak Powet in Element / 1,,, 1,,7 1,,,

AvetoDe Powet in Core Fuel ' '

I I 6.6 8.8 8.5 5.6 -

@ I 33 1  % 55 2.13 2.65 2.52 1.85 ,5 6.0 8.1 9.2 7.5 1 1 I U 1.60 2.46 2.53 2.28 l_I I I I 6.1 6.7 6.6 4.2 I is 1 - . .

fiTl I as l I 461 ,

1.80 -- 2.13 2.13 1.56 O.8 O.23 l LEU CORE Power / Element. 5.4 6.1 4.9

• rnlillWe tte Peak Power in Elementt  ! 32 I b { 6d !

1.94 1.99 1.80 AveroDe Power in Core fuel '

'_l I

6.7 8.7 8.0 5.1 122 l I as l l 44 i Firl 2.24 2.58 2.62. 1.73 5.4 8.5 9.1 6.9 112 l [ 23 l 134 l l 45 ]

1.60 2.56 2.59 2.21 l ._3 I 5~ l 5.8 7,1 6.3 l ia l _.

124 l 1351 ,

2.02 -

2.20 2.16 4.2 1.0 14G l I 26 l 1.61 0.26

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Tabic 1. Sensitivity of LEU Core to the Ntunber of ~ '

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Fueled Cyllnders in the Pm tlal Element - g tg 4

Fuelin LEU . Reactivity Cliange, jf

.Caan Partial Element  % 6Mt !6 1

2-Cylinder 2 Only No Partial Element 0.0. (Reference Core) -

0.33

$4 3- Cylinder 4 Only 0.14 ; F 4 Cylinder 6 Only 0.26. ,P Cylinders 2 and 4i 5

0.37 6

7 Cylinders 2 and 6 Cylinders 4 and 6 0.47:

0.57 '

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8 Cylinders 2,4, and 6 0.72-

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-i Tobic E, Removal and Replacement Scouence of IIEU and LEU Fuel Elements--

llEU Fuct  : LEU Fucl .

No. of Console-Fucl Elcinent No. Removal Order . Insertion Order. . Meter Readings 25 (partial) i 16- 1st.

46 2 tcplaced.by # 14 14 cmpty

.i-15 < t. ..

4 13 3 14 '2nd 12 4 '13

>f 22 5 12 32 ,6 11: '3rd- .-

43 7 ~10

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24 11 6

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( MECllANICAL ENGINEERING DEPARTMENT

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,1 REACTOR PERIOD AND REACTIVITY a

Nf Ob_le ctive Q

a tj To determine the reactivity worth of a portion of the stainless I fj steel regulating rod.

,1 .

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' References

,(

ij " Introduction to Nuclear Reactor Theory" by John R. Lamarr.h.

',( pp. 420- 428, 43*l. 4 39 and 4 41-4 42.

.h '

" Nuclear Reactor Physien" by Rayrnond L. Murray, pp. 156 160.

3

" Intr oduction to Nuclear Engineering" (Second edition) by

( 9 Raymond L. Mur ray, pp. 131-138,

'5 Theory M The theory involved in thle experiment in explained adequately in

?! the three references given above. The inhour equation is given on the

'" accompanying pagen, u

Procedure 4 n q

1, The reactor will be made critical with the Reg. Rod about 89%

withdrawn and the piconmmeter reading about 2 on the upper scale (j,

when the scale selector la set for 3 X 10-8 amperes. ,dbtain

"' Introduce a step 6k by moving the Reg. Rod to 100%. .

j doubling times by clocki"8 the time elapsed for the needle to move from 3 to 6 and f rorn 4 to 8 on the upper scale. As noon

')

3 as you have taken the reading at 8 ewitch the scale selector to 10 X 10*0 amperen in order to avoid an involuntary scram, s,

2. The reactor will now be made critical with the Reg. Rod about '

84% withdrawn and the pleoammeter reading about 1. $ on the e upper scale when the scale selector is set for 3 X 10-8 amperes.

il 1 Repeat the remainder of step # 1.

f ln 3

f 1

- - - . , . , , . . ~ -

_.__..,.,.__..._g e; . ..* .';-:4 Q

l%. .'

n , ,

. t[ * ',

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v.

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Procedure (continued) e

3. The reactor will now be tnade critical with the Reg. Itod about .

075% withdrawn and the picommmeter reading about 2 on the 10 X 10*9 ampere scale. Introduce n step f k by moving the_

P e

Reg. Itod to.100% Turn the scale. selector to 3 X 10-8 amperes when the needir reaches 7 on the.10 X 10-9 ampere scale. TfL the "Up'lamit" switch clicks-and the "Up Limit" light goes on. - '

follow the remaind er of the procedure in step # 1. Otherwise -

it may be necessary to take the first doubling time from 3.5 to 7 or to omit it altogether.

Ilesulta llequl'red in each of the three steps of the' Procedure. average the two doubling timen and determine the reactor period. From thu accompanying

, tables determine the reactivity input for each step in terms of both per cent and cents.

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ATTACHMENT II (

EXPE ilMENT d20 - POWER LEVEL DETERMINATION b

Introduction:

In high power reactors, it is possible to determine a given power level by  !

calorimetric methods; core and coolant temperature measurements, coolant flow rates, etc. Low power reactors do not have this advantage since the heat generated during normal operation is small. Therefore, the approach used to measure a given -

power level of the Manhattan ZPR will not involve the product of nuclear fission, heat,'

but the initiator of the reaction, neutrons. Each fission which occurs in the reactor '

core releases an average energy of 200 mev. The rate at which energy is being

' released, the power level, is proportional to the itssion rate. A measurement of the fission rate will therefore constitute a power level determination.

Two methods will be employed to determine the power level of the reactor and thereby calibrate the Log N and Linear channels. The first method will be a suberitical one while the second will employ the use of gold foil as an activation detector. An absolute thermal flux measurement will be made at the core center with-the standard gold foil. This foil will be counted on an end-window G-M counter whoc -

officiency for the standard gold foil has been determined from a previous standar; pile irradiation. The results of the measurement will yield the average therr..d ,

flux in the reactor core. A cadmium ratio measurement of the gold foil will also be made in order to determine the fraction of the total fission rate due to epithermal neutrons.

Theory:

Assuming an all thermal homogeneous reactor model, the following ex-pression describes the total fission rate occurring when the reactor is in a steady state condition:

Rf *Et ofyol, # dV (1) where c = the thermal neutron flux, the total neutron density times the most probable value of the thermal neutron velocity distribution at standard temperature.

If = macroscopic fission cross section of the fuel at the most probable velocity corrected .if necessary for a non-1/v behavior.

The neutron flux will have a certain spatial distribution in the core depending on ,

the geometry of the system. If the average thermal neutron flux can be determined, equation (20.1) will simplify to N i" i T Y score 20-1 ,

e 6*'7'"*****"*O'",. ,h _.m, . _. ,

___-.__..._,y e .. .- .

v y

^

7 k

with 4 = f v0 dV/ J vd V , the spatial averago of the thermal flux. [

f i The power level of the reactor corresponding to the average thermal flux is then equal to the total fission rate dis ided by the number of fissions per second required l to produce one watt of power. 1:or a reactor utilizing U-235 fuel, POWOP " I = ---

g W atts (2) 3.1 x 10 3.1 x 10 l

Although equation (20. 2) has been derived for a simplified model, its g accuracy when used for the Manhattan ZPR power calibration will depend primarily B on the average thermal flux measurement and to a lesser extent on a correction ,

made for non-thermal fission. The energy distribution of the neutrons in the core ,

actually covers a wide range, from fission energies down to the thermal region. '

Defore the neutrons are slowed down into the thermal region, some will be '

1 captured in U-235 and U-238 and hence cause additional fissions. The fission occurring in U-238 is caused only by neutrons of high energy. The threshold I reaction's contributions to the total fission rate can be assumed small for the Manhattan ZPR reactor since its moderator to uranium volume ratio is appreciable I and its fuel is enriched with the U-235 isotopo. Very fast fission is nornmily ac-counted for in the four factor formula by the factor c the number of neutrons pro-duced by all fissions divided by the number produced by thermal fission. In the I Manhattan ZPR non-thermal fission is predominately resonance fission since U-235 1 has finite fission cross sections at all energies. The annount of epithermal fission  :

a can be determined by a simple cadmium ratio measurement of Manhattan ZPR type  ;

fuel. The fission product activity of a bare and cadmium covered fuel sample can be counted on a proportional counter after two similar irradiations in the reactor j core. Their ratio will yield the amount of non-thermal fission to the total fission after proper corrections for sample. weight differences, irradiation times and power level differences have been made. The final power level expression then be-i cotnes l EoV I Clt W atts p, -

CR-1 (3) 3.1 x10* -

- fuel where Clt a fuel sample's cadmium ratio.

The average spatial thermal flux is actually the total neutron density times a velocity of 2200 m/sec. This is true because the absolute thermal flux is mensured

.vith a 1/v absorber using its cross section at this standard velocity. The macro-neopic Ilssion cross section of U-235 is that of the same velocity. Actually, since this cross section does not have a 1/v dependence, a correction can be made to make it an equivalent 1/v cross section, which will give the correct fission rate in aquation (20. 3). This correction is based on a Maxwellian neutron distribution being 20-2

.--_.._-.....-.__.--------====N1

_. ._.... _ = _ _ - ~ _ _ _ -

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pres:nt cnd is a function of the neutron temperature. The temperature of the neutrons in the Manhattan ZPR core is approximately "K which correspondu to i i

a correction factor of 0.

i The absolute flux measurements at the core center with a standard gold I foil requires the knowledge of the fundamental activation analysis. The activity  ;

l of the foil immediately af ter the irradiation can be expressed as s

Ag =N c T "act E (I ' " ) (4) where NT = total number of detector atoms. I;

=

[

"act. thermal neutron activation cross section at 2200 meters /sec. [,

W

thermal neutron flux (nv2200) i K

thermal flux depression, self shleiding factor A = detector's decay constant t a irradiation time.

i If the detector is counted on a system whose inverse officiency, E(dis /nec/

count /sec), is known at a time t after irradiation, the detector's activity becomna g -

EC = Ao e'AI c (5) where C = counts /sec at (c '

Solving (20. 4) and (20. 5) for the neutron flux ECe Xt e O * (0)  ;

"T act K II ~ " A t

'The resulting neutron flux is the proper flux to be used in equation (20. 3) only if tho activity is caused by thermal neutrons alone. Since the cross section of gold has a strong resonance activation peak, a correction must be made to separate ths resonance activation from the thermal activation. This can be done by making a cadmium ratio measurement with the standard gold foil. If CR is the measured value, the thermal activity of the detector becomes A CR _1_

o th = A o CR A oFoll 20-3 '

- +

.. c. . : ~" ~m  :'  :' 2 ' ~rm_? n ~ N

m __ _ _ _ .

.- _. , , }

o II:nce, the flux, in(20. 6) mual be multiplied by the factor in the brackets to yleid

- y the correct thermal flux.

Procedure:

Method 1 of determining the core power level will be a suberitical method.

L Upon completing the initial critical for the Manhattan ZPR, the regulating rod will be calibrated using period measurements. Af ter determining the worth of the regu- b 4

lating rod, the core will be held at criticality with the source in and all indicating instrumentation readings will be recorded. Af ter all readings at criticality are e completed, the calibrated regulating rod is inserted until the reactor is sub- 1 critical by a known amount. The reactor power level will decrease and level at a new lower level which can be calculated using the following relationship: ,,

c,

0. 2 So ,

Power (watts) = 10  ;

(1-Keff) 7. 55 x 10 0

S is estimated from the Manhattan ZPR source strength (1. 0 x 10 n/sec)  ?

g ,

c k ggg is known from the rod calibration.

Method 11 of determining the core power level will be done using gohl foil I -

trradiation data. A bare gold foil and a cadmium covered gold foil will be placed i on an aluminum holder so that each foil will be located a distance 0. 5 in, above 2 and below the center of the active core length in fuel element #20 in core position

33. This distance will place the two foils at the position calculated to see the ,

average axini flux in this fuel element. The max / avg. of the core radially will l b3 1. 44. t Place the foil holder into fuel element #20 between fuel plates #3 and #4 and '

bring the reactor to critical leveling at same instrument readings used for Method 1.

Ramove the start-up source maintaining criticality. Irradiate the foils for 15 [

minutes. After 15 minutes irradiation, shut down reactor and remove foils. The  ;

foils will be counted on the Manhattan G-M Counter that has been standardized for the gold foil. Count foils and correct count rate measured back to zero time, i, l

The average thermal flux seen by the gold foil is determined as follows:

i A" AU)Bd (CdR-1) _ At.Wt. e Xtd  ;

p , ,_

th y Au Cdn 0.623 1e -Ate j

act 20-4 t

, y~{

. l- , c.

g = absolute activity per gram of bare foil at time t sec gm) d CU d A A. = . , -

O x 10 sec cd

,n et = 85 barns, t

d U "I " * * " # *E " * " " "" """"E '""

i g = duration time of exposure (sec)

The power level of the core is calculated using the following relationship:

~II Power (watts) = 3. 54 x 10 xMx 5 core Th where M = Mass of U-235 in core at time gold foils are irradiated in gms.

Au .

(A core = ,o x 0,695 th P

T p = Ratio of total power to power resulting only from thermal fissionn - 1. 20 Th Prerequisiten A. Normal start-up instrumentation must be operative B. Initial critical loading must have been completed C. Equipment required:

1. Two gold foils with known waights

2. One cadmium cover

3. Foil holders i

4. Manhattan counting room equipment

5. Calibration data for one safety rod Precautions:

Normal operating procedures will be followed. Power level must be held standy during irradiation of gold foils, and as low as practical for the performance '

of the experiment. A power level of abundred mW 's desirable. Foll holder will  !

be placed into fuel element #1 holddown rod with core in shutdown condition.

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20-5

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.- 'llencNic Conditlons_: -4 .

+4 The reactor during this measurement will have the initial critical loadin:. -;'i(

- All instrthnentation will be operative s; o ,

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'~. . DATA ' SHEET 1 _4, 6

• j INITI AL POWER LEVEL DETERMINATION s i

METHOD l' 4

J Core Loading # ._

i 4

U-235 Contoni _ '

c Pool temperature - " {

Location  ;

Strength _

l Start-up Source - Type _

i Critical Rod Positions _(Source in core) t

, 4 r

. Reg. Rod  ;

' Shim Rod t.

~ Critical Rod Positions (Source removed)  ;

1. 5-  :

1. 2 #3 #4 Reg. Rod #1__ i

1. 2 # 3- _#4 #5 _ ,_ i .

. Shim Rod #1  :

Trial #1 Trial #2 '

. Period Measurement

- Reg. Rod- Critical (source out)

! . Position Super critical (source out). _

L t

t Trial #1 Period see l

Trial #2 Period sac l'

' Use in-hour curve to determine 4 k required to give above periods Differential worth of heg. Rod _ Trial #1 Trial #2 _,

insert start-up source (note core must be made sulicritical before inserting sourcel

[ Return reactor to critical, adjust all rods in positions of previous critical with

! source in, t

I-

, 20-?

U m.,,. ., .4

- _.a_,maa,s._ .. g ,j.

m .

7 . -.

Record all instrument readings at criticality i.

R \

Linear 06 Range g _ Start-up channel

+:

g Start-up channel Log N ,

i g -Inse ?t Reg. Rod until tN nd. f c re is suberitical a known amount. ll (use whove differentlai worth to determine Kgff)

Keff (Subcrittent ]

j Reg. Rod iJ Allow power level to drop until it levels off. L After a steady state subcritical condition. i

.is reached; record all instrument readings.

1 Record all instrument readings at level subcritical position i

Start-up channel #1 Linear  % Range j

. .)

Start-up channel #2 Log N .{

1

)

'1

^k

,i 1

l 1

l

- i 1

l 20-e l

. ur : 2.r . . r-> . . .t . .. - ~ . -. ... , __

I ;f DATA SHEET 1 - 5  ;,

INITIAL POWER LEVEL DETERMINATION >

\;

METHOD lli t

Core Loading # Date -

t U-235 Content ., - ll Pool Temp. E ll E Start-up Source Type Strength- Location j

]d Prepare foil holder with two gold foils so that when placed in core the gold foils will- Lj!

.be located 9. 5 inches above and below the core midplane. .One gold foil will be cadmium: l covered and the other will be bare...  ;

i 70,11 Data i Gold Foil #1 (Bare) Gold Foil #2 (cd covered) ,

.id #

-j w t.' _

[

MNith the core shut down, insert the foil holder containing the foils into the hold down red of the fuel element in #33 position.-  :

scing the core to criticality and allow power to rise one decade above previous (

criticality instrument readings. Level the power and remove the start-up source. 1 M

Record time Power Level attained j Critical Rod Positions (source out) )

R eg. R cd

]

- Shim Red

'nstrument Readings i 3

]r Start-up Channel # Log N kart-up Channel # Linear  % Range.

d 20-9 i f

. _ _ _ . - . _ - . _ . _ _ . _ _ - - - , - - - - - - - - - - - - - - ~ - - - - - -

Wm_ ~ ~ -

4q

.- ,e ~*

T

'l .f. .

b e . .

Af ter ~15 minutes shut down core and remove foil holder .];

M.  ?,

. Time of shut down fl i ..

l-exposure = te = seconds .

h,

Foll Data l.

%?

' C it Gold Foil #1 Gold Foil #2 1

. Time Counted.

p td (decay time)  ;

Total counts -

c Activity (dis /sec. gm)

Counter efficiency Cd Ratio = A Bare *

. A Cd covered ...

4. 7 e

rin ii h l j

d,e a t

.<l e

' 20-10

.=_ _h__ _ ________________._ _ ____ _ . ___

EXpERI ENT #27- - ROD WORTH MEASUREMENTS .  ! )y

.i L,Q Test Pudose:

!]

To determine reactivity worths of the cadmium control rod and the stainlenn !9 steel regulating rod.

![

iy rest Summary: -

ii

!O Rod worths will be determined using the rod drop and positive period {y nothods. By reference to the' Core Diagram, the regulating rod is between core positions 3 0

.3, 23 and 24, while the cadmium safety rod is located between core positions 32, 33 and: _  :

L3.

~ [l

)rerequisite Operaticus: lj i:

1 f.

d General: -

1) 4 .i

' n The reactor core loading shall be limited to the initial critical loading  ; ,

or the start of these tests.

li l.

I Instrumentation:

g. _

A11: instrumentation shall be checked out and in proper working order. J

)' recheck shall be mado of the initial criticality data for the rods 50% withdrawn. L ,

q; recautions: i Rod Withdrawal i O

Do not introduce a period less than 20 seconds.

[ Dypass Switches 7 l :-

l- Bypass the same scram circuits as for the " Criticality Test." d V1 l1tial Plant Conditions:

1 l'

t -

E 1. .H20 at normal operating level, jj* :

?

c .

2. Initial criticality core loading.

j p

.3.. Source in position as indicated in Core Diagram. , i is l l 4.s Reactor ventilating system operating.  !!

- 5. Reactor water purificating system operable.

l f!

ti l

-l )

[:

l Il t

't !

l 27-1 ~

,~ . . . ~__. .._._.._-Z

~

~-l ~ W ~~ ;.

e ti

proclduram y i d

. Rod Drop Method '

n a

1. Using the start-up channel, remove source and measure any extraneous i!

background level. a p

2. Insert the source to the original position. N n

3. Bring the reactor to critical and adjust the neutron flux level to a p decade'or two above the neutron flux level due to the source with all rods in, and opera +e for a few minutes to allow most of the delayed neutrons to stabilize.

j'a a

4. Remove the source.

(

5. Record the rod positions as show'n on the fine position indicator.'

O. Take three one (1) minute counts on the start-up channel counting I circuit. Record these counts. *

7. Drop both rods by pressing manual scram button after operation for. ,

10 minutes.

8. Measure total counts on the start-up channel between 30 and_00 seconds after the instant of scram. This operation may be done manually by use of a stop wutch but it is recommended that an automatic timer be employed to start and stop the BF channel scaler for the desired time interval, 3-

9. Repeat steps (1) through (6) except position rod to be tested full o'ut.

10. Drop the desired rod by reducing the magnet current. .

11. Repeat step (8),

12. Repeat steps (9) through (12) for the second rod. I l 13. Calculate the following ratio for each rod tested.

l R = (Count Rate at Power (cps) )

Integral Counts (30 - 90s) i

14. ' Determine the worth of the rod tested by reading the~ negative reactivity effect, i.e. % excess k, from the graph of do excess k versus R using the measured  !

-value of R. (see accompanying graph).

27-2

e r

.- !gl Positive Period Method be

1. Bring the reactor critical with the neutron flux level a few decades nbm c '

the shutdown source level. Withdraw the shim safety rod fully while keeping the flux i l level stabilized by adjusting the regulating rod critical position. -

2. With the flux level stabilized, withdraw the regulating rod a pre - l l determined distance from its critical position and measure the resultant period on the i Linear and Log-N circuits. Period measurements should be between 30 and 50 seconds '

in order to obtain reasonable data. Record initial and final positions of the regulating rod.

l. l

3. Insert the shim safety rod to stabilize the flux at the level measured 1 l

4 prior to the period test. 1

4. Repeat steps (2) and (3) until regulating rod is calibrated over its l complete range.

5. From the period measurements and the reactivity versus period curve, .

I plot the worth of the regulating rod versus the distance above the lowest critical level. l

6. Repeat steps (1) through (5) for the shim safety rod.

NOTE: Since either rod can shut the reactor down, shim i and regulat_i lng can be interchanged, l}ata Required: j

1. Differential and integral rod worth curves for each rod as a function of rod position.

2. Rod worth measurement data for rod drop method. ,

3. Core conditions under which measurements were made.

n 1 1

1 27-3

)

s _ .. _ ..._......_ _ . _ . . . . . . . . . .. ___J

--... o

.= _,.

1 Attachment 111 Prooosed Revisions To Technical Soecifientions-The following proposed revisions to the Technical Specifications are divided into-- .

two (2) groupings; Part A are revisions to correct existing terminology; & Part B are l the revisions to accommodate the HEU to LEU conversion l A Corrections to Existing Technical Specifications: j (1) Page 11, delaved neutron fraction Revise to read "when converting between absolute and dollar value >

reactivity units, a beta effective delayed neutron fraction' of 0.0078 is - ,

used".-

Rationale: A delayed neutron fraction of 0.00645 for converting between absolute- 4 and dollar value reactivity units is .not correct._ A delayed neutron ,

fraction of 0.00645 is the value for pure U 235, 11 is the number of; delayed neutrons divided by the total number of neutrons emitted _when an. i atom of U 235 is fissioned and is independent of the uranium enrichment ,

and the geometry of the reactor. The correct conversion unit is the '

effective delayed neutron fraction, which depends on the reactor geometry 4 and the diffusion properties of the medium (see ANL 5800,2nd Edition,- a Reactor Physics Constants, July 1963, pp. 441444). .- A. beta effective  ;

of 0.0078 for both the HEU and LEU cases has been computed and .

suggested by the Argonne National Laboratory.

(2) Page 12, reactivity limits l Delete the sentence "For the MCZPR the reactivity _ limits are 0.44%

AK /K (0.68$) at 110.60 F".

j ,

Rationale: The reactivity limit is specified in paragraph 3.1.3 on page 31. - Since l Individual reactor parameters- are not generally included. In the l definition, and the sentence is repetitious of the paragraph 3.1.3 wording, L

consistency would suggest it be deleted.

t (3) Page 31, 3.1.3 Specifications

l. A Change (0.68$) to (0.56$)

! -- B. Change (0.72$) to (0.59$)

l Page 3-1, 3.1.4 Bases l Change (0.68Si to (0.56$)  :)

Page 3 2, 3.2.3 Specifications B. Change (1.40$) to (1.15$), and (3.88$) to-(3.21$) '

D. Change (0.154$) to (0.128$)

Rationale: The proposed revisions are consistent with using a beta effective of ,

0.0078 to convert absolute to dollar value reactivity units, f

l f:

(-

l

~

a

//l d

-* -g B. Proposed' Revisions to Technical Specifications to Accommodate' HEU to' LEU Conversion t

(1) Page 21, 2.1.3 Specifications t

Revise to read: "The safety limit shall be on the temperature of.the fuel element  !

cladding, which shall be less than 10800 F".

Rationale. The cladding of the current HEU fuel elements is composed of 1100  :

(or 2S)' aluminum which has a melting temperature of 12200F. -The cladding of the new LEU fuel elements will be 6061 aluminum, which ,

has a melting temperature of 10800 F.

(2) Page 21, 2.1.4 Bases ,

t Revise to Read: "The melting temperature of the aluminum used as cladding on the fuel' >-

elements is 10800F. _Therefore,,In order to maintain fuel element Integrity, the cladding temperature must not_ exceed 10800F, As reported in Section 6.1.2. of " Analyses for- Conversion- of the ,

Manhattan College Zero Power React'or From HEU to LEU Fuel" by J.

Matos, and K. Freese of Argonne National Laboratory (Reference 1).

The maximum cladding temperature that can ever be reached is only 2390F (11500) and reaches this level only during the Maximum-Hypothetical Accident. The specification - therefore,~ provides-assurance on the integrity of the fuel within the cladding".

Comment: '

Based on the same reason as in (1) above (3) Page 2 2, 2.2.4 Bases-Revise to Read "Since there is no forced circulation cooling, the reactor core is cooled' '

by the water surrounding the reactor core. Therefore, the only

~

parameter which could.be used as a limit; for'the' fuel cladding:

, temperature is the reactor power. The analysis in Reference 1 shows i

that even for the Maximum _ Hypothetical Accident (a reactor power excursion of 183 Kilowatts), the maximum cladding _ temperature Ll reaches only 2390F (11500). This temperature islmuch lower that the temperature (10800F) at which cladding damage could occur. 3

Therefore, a large safety margin exists between the safety system set-

1. point and the cladding safety limit ,

l 1 l The revised parameters are based on the- ANL accident analysis (Reference 1) (Note in Section 6.1.2 (p.23) of. Reference.1 that ANL.

personnel computed a peak power of 221 kilowatts and a peak cladding

, temperature of' 2410F (11600) instead of the peak power of 147 kilowatts and a peak cladding temperature of 2210F that is quoted for the HEU core in our Technical Specifications and SAR. None of our conclusions for the HEU core change because of this difference). 1 1 j i

l 1

l l ,

., J. i

$ . . i

-(4) Page 31,3.1.3 Specifications, Paragraphs A and B., and 3.1.4 Bases 1 1

Comment: The reactivity limit for the LEU case has been acsumed to be the same l 0 0.44 % AK/K as for the HEU case. If the 1966 isothermal heating. l l

experiment were redone for the LEU case, the maximum excess - +

reactivity may not be exactly 0.44% AK/K at a pool water temperature of exactly 110.60F.

l L With regards Paragraph B, i l- The actual minimum shutdown rnargin of the LEU core will not be l

known untlL the reactivity worth of the regulating rod and the maximum excess reactivity are actually measured, in Section 5.4.4 of Reference 1- the minimum shutdown margin in the LEU core is.

estimated to be 0.56% AK/K larger than in the HEU core because the fuel element in position 46 of the HEU core was moved to position 14 of the LEU core in . order to increase the reactivity worth of the regulating rod.

After completion of testing of the LEULcore, revisions to-the paramters in -

Paragraphs A&B will be proposed, as well as_ the parameters in the first sentence of ,

section 3.1.4.  :

j (5) Page 5 3, 5.3.2 Reactor Fuel _;

Revise the first four sentences to read: ,

"The fuel portion of the elements consists of six concentric cylinders -

formed by mechanically Joining and positioning eighteen curved fuel; plates wihin grooves of three spacer webs. The cylindrical fuel plate: i consists of 0.020 inch thick ' U3 Sl2 - Al fuel _~ meat containing.

uranium enriched to 19.7510.2% in U 235. and clad on both s! des ..

with 0.015 Inch of aluminum, making the total fuel plate thickness- j 0.05 inch. The nominal U 235 content of each full fuel element is 235 grams. The inner diametar of the innermost cylinder is~ about 1.25 inches and the spacing between' adjacent. cylinders (water channel width) is 0.118 inch".  ;

Rationale: The revisions are consistent with the parameters of the new LEU case. ,

The last 2 sentences relating to the exact number of fuel elements in the LEU case will be revised after'a critical core satisfying all the.. '

Limiting Conditions for Operation is assembled.

i In.this regard, the nine fuel plates in cylinders 2,4, and 6 of the -

partial fuel element are removable. The nominal U 235 content'of ,

the three fuel plates in each of the three full cylinder are 27.4 43.7, J and 58.4 grams, respectively (see Reference 1, Table 1, p.5). If the partial fuel element is needed, the minimum loading Is 9.1 grams U-235 in one fuel plate in cylinder 2 and the maximum loading-is 129.5 grams.

___m__._..__.m_ . _ _ _ _ _ _ _ _ ,